The following relates to the nuclear reactor arts, nuclear reactor operating arts, nuclear power generation arts, nuclear reactor safety arts, and related arts.
In a pressurized water type reactor (PWR), a nuclear reactor core comprising fissile material, e.g. 235U, is disposed within a pressure vessel and immersed in primary coolant, usually water. The primary coolant flows upwardly through the reactor core and is heated by the radioactive core. The primary coolant flows through a steam generator where it heats secondary coolant water to convert the secondary coolant to steam, which is used to perform useful work such as driving a turbine in the case of a nuclear power plant. An advantage of PWR designs over some other system such as boiling water reactor (BWR) systems is that the secondary coolant does not come into contact with the nuclear reactor core. Conventionally, the steam generator is separate from the PWR and a primary coolant circuit conducts primary coolant between the PWR pressure vessel and the external steam generator. This primary coolant circuit introduces large-diameter piping and hence is a potential location for a loss of coolant accident (LOCA). In some PWR designs, the steam generator is disposed inside the pressure vessel (sometimes referred to as an “integral PWR”). An example of a deployed integral PWR is the Consolidated Nuclear Steam Generator (CNSG) system developed by Babcock & Wilcox and employed in the German nuclear-powered ship N. S. Otto Hahn which was in commercial service between 1970 and 1978.
A loss of coolant accident, i.e. LOCA, occurs when there is a substantial interruption of the primary coolant circuit, typically through a pipe break at a vessel penetration into or out of the nuclear reactor pressure vessel. Besides a LOCA, a nuclear power plant can experience other types of abnormal operating events, such as a station blackout or a loss of feedwater event. A station blackout occurs when external power to the nuclear island is interrupted. Although a nuclear power plant generates electricity, it normally relies upon the local power grid for electrical power to operate equipment such as pumps, cold water circulation systems, and so forth. A loss of feedwater event occurs when the secondary coolant flow is interrupted, either through a pipe break or through an event, such as a turbine trip, that causes safety valves to interrupt the secondary coolant circulation. As reactor heat sinking is provided by heat transfer from primary coolant to secondary coolant in the steam generator, a loss of feedwater event is effectively a loss of heat sinking event.
The safety systems of a nuclear power plant are extensive, and include (in addition to the pressure vessel of the nuclear reactor) a containment structure surrounding the nuclear reactor, typically made of concrete, steel, or steel-reinforced concrete, and an emergency core cooling system (ECC) that is designed to depressurize the pressure vessel and containment structure, and to transfer heat from inside containment to an ultimate heat sink (UHS) comprising a body of water located outside of containment. In a typical ECC response, any overpressure inside the reactor pressure vessel is vented into the containment structure, borated water under high pressure is injected into the pressure vessel, water is poured down the exterior of the pressure vessel and drains into a flood well at the bottom of the containment structure, and condenser systems condense the steam and reject the latent heat to the UHS pool. The borated water serves as a neutron poison and, together with scram of the shutdown rods, quickly extinguishes the nuclear chain reaction. However, residual decay heat from short half-life intermediate products of the nuclear chain reaction continue to generate decay heat in the reactor core, and the heat output of the core decays exponentially. This decay heat is initially expelled to the UHS pool by the ECC condensers; after depressurization, low pressure heat exchangers take over to continue to reject decay heat to the UHS pool.
In a LOCA, primary coolant in the subcooled state flashes to steam and escapes into containment where it is condensed by the ECC condensers. In a station blackout or loss of heat sinking event, temperature and pressure may rise inside the pressure vessel due to interruption of primary coolant circulation (e.g., due to shutdown of the reactor coolant pumps in a station blackout) and/or due to interruption of the heat sinking (in the case of a loss of feedwater event), and if the pressure in the pressure vessel becomes too high then relief valves vent excess steam to containment (e.g., into a refueling water storage tank, RWST, located inside containment) and the ECC condensers accommodate any pressure rise inside the containment structure.
All these are abnormal events, and require extensive post-event actions, e.g. removal of radioactive primary coolant water from the containment structure, filtering of the (remaining) primary coolant water inside the pressure vessel to remove excess soluble boron compounds, regeneration or replacement of ECC condensers or other ECC components, replacement of purified water in the RWST, replenishment of the UHS pool, and so forth, before the nuclear reactor can be restarted and put back into service. Additionally, any event in which primary coolant water escapes into the containment structure (even via a designed pressure relief valve) is an event in which radioactive primary coolant has reached the “secondary” containment level provided by the containment structure.